The present invention relates to methods of preparing medically useful radioisotopes, particularly high specific activity no-carrier-added radioisotopes, and more particularly to methods of preparing high specific activity, no-carrier-added Lutetium-177 (177Lu).
The use of beta particle-emitting radioisotopes for applications in nuclear medicine, oncology and interventional cardiology is rapidly increasing, because of the availability of new pharmaceutical targeting approaches, which effectively concentrate or localize the radioactive vector at the target site with low uptake in non-target tissues. In this manner the energy released from radioactive decay can be localized for killing cells at the target site, such as a tumor. In this regard the use of such radiopharmaceuticals has been shown to be effective in treating a variety of tumors.
Some peptides radiolabeled with 177Lu (T xc2xd=6.7 days), a low-energy beta emitter (Emax 0.497 MeV), are considered particularly useful as a result of rapid cellular uptake whereby radioactive decay occurs within the cell. Low-energy beta emissions are highly effective in the immediate vicinity of the cell, and the effect on adjacent, normal and sensitive tissues is minimal. Therefore, it is desirable to obtain 177Lu in high purity, high specific activity form.
177Lu can be produced in a nuclear reactor by the conventional xe2x80x9cdirectxe2x80x9d production method involving neutron capture of enriched 176Lu, as shown in FIG. 1. Since the nonradioactive target atoms and radioactive product atoms cannot be separated by chemical means, the radioactive 177Lu is diluted with significant amounts of the 176Lu carrier. Moreover, metastable 177mLu is also produced by the xe2x80x9cdirectxe2x80x9d method. Metastable 177mLu, having a half-life of 160 days, is generally considered harmful for nuclear medicine applications because of the potential for extended patient exposure to radiation.
The amount of 176Lu target (carrier) determines the specific activity (Curies/gram), with higher specific activity products being produced in higher neutron flux reactors such as the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). Although high specific activity ( greater than 80 Curies/mg) can be produced by this method in a high flux reactor such as the ORNL HFIR, it is desirable to obtain higher specific activity 177Lu in order to attain a higher specific dosage and also to extend the shelf life of 177Lu inventory. It is also desirable to have a method available to provide high specific activity 177Lu that is essentially free of long-lived 177mLu impurity.
There has been interest for several years in the xe2x80x9cindirectxe2x80x9d method of reactor production of 177Lu, which is obtained from decay of the short-lived (T xc2xd=2 hours) 177Yb radioisotope, which is produced in a reactor by irradiation of enriched 176Yb targets (also shown in FIG. 1). The indirect method is advantageous over the direct method in that there is little or no metastable 177mLu produced by the indirect method. However, there remain large amounts of Yb target material that should be removed.
A major hindrance in the feasibility of producing no-carrier-added 177Lu is lack of an effective method of the separating no-carrier-added 177Lu from high levels of Yb. Separation of adjacent lanthanides is notoriously difficult using conventional ion exchange chromatography and other known methods because of the close similarity in the chemical properties of lanthanides.
An effective and efficient method of separating 177Lu from reactor-produced 177Yb is needed to provide high specific activity, no carrier-added 177Lu. The only successful preparative scale separation of Lu and Yb which has been reported in the literature is the recent paper by Lebedev, et al., which describes the use of a difficult and cumbersome cementation process, in which tracer levels of 177Lu are separated from macroscopic levels of Yb by the repetitive, selective extractions of Yb using a sodium (mercury) amalgam from chloride/acetate electrolytes, followed by a final cation exchange purification step. Such a method is not generally considered feasible for preparing 177Lu in sufficient amounts for practical applications in nuclear medicine because of potential contamination with toxic mercury.
Accordingly, objectives of the present invention include provision of: methods of separating microscopic amounts of Lu from macroscopic amounts of Yb; methods of preparing 177Lu that is at least one of high specific activity, no-carrier-added, and essentially free of 177mLu; and methods of treating disease using no-carrier-added 177Lu. Further and other objectives of the present invention will become apparent from the description contained herein.
In accordance with one aspect of the present invention, the foregoing and other objects are achieved by a method of separating lutetium from a solution containing Lu and Yb includes the steps of: providing a chromatographic separation apparatus containing LN resin; loading the apparatus with a solution containing Lu and Yb; and eluting the apparatus to chromatographically separate the Lu and the Yb.
In accordance with another aspect of the present invention, a composition of matter comprising essentially 177Lu which is characterized by at least one of: essentially 177mLu-free (no or insignificant presence thereof), no-carrier-added, and a specific activity of at least 100 Ci/mg Lu.